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Advanced neutron source reactor thermal-hydraulic test loop facility description [electronic resource].

Published
Washington, D.C. : United States. Dept. of Energy, 1994.
Oak Ridge, Tenn. : Distributed by the Office of Scientific and Technical Information, U.S. Dept. of Energy.
Physical Description
26 pages : digital, PDF file
Additional Creators
Oak Ridge National Laboratory, United States. Department of Energy, and United States. Department of Energy. Office of Scientific and Technical Information
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Summary
The Thermal-Hydraulic Test Loop (THTL) is a facility for experiments constructed to support the development of the Advanced Neutron Source Reactor (ANSR) at Oak Ridge National Laboratory. The ANSR is both cooled and moderated by heavy water and uses uranium silicide fuel. The core is composed of two coaxial fuel-element annuli, each of different diameter. There are 684 parallel aluminum-clad fuel plates (252 in the inner-lower core and 432 in the outer-upper core) arranged in an involute geometry that effectively creates an array of thin rectangular flow channels. Both the fuel plates and the coolant channels are 1.27 mm thick, with a span of 87 mm (lower core), 70 mm (upper core), and 507-mm heated length. The coolant flows vertically upwards at a mass flux of 27 Mg/m²s (inlet velocity of 25 m/s) with an inlet temperature of 45°C and inlet pressure of 3.2 MPa. The average and peak heat fluxes are approximately 6 and 12 MW/m², respectively. The availability of experimental data for both flow excursion (FE) and true critical heat flux (CHF) at the conditions applicable to the ANSR is very limited. The THTL was designed and built to simulate a full-length coolant subchannel of the core, allowing experimental determination of thermal limits under the expected ANSR thermal-hydraulic conditions. For these experimental studies, the involute-shaped fuel plates of the ANSR core with the narrow 1.27-mm flow gap are represented by a narrow rectangular channel. Tests in the THTL will provide both single- and two-phase thermal-hydraulic information. The specific phenomena that are to be examined are (1) single-phase heat-transfer coefficients and friction factors, (2) the point of incipient boiling, (3) nucleate boiling heat-transfer coefficients, (4) two-phase pressure-drop characteristics in the nucleate boiling regime, (5) flow instability limits, and (6) CHF limits.
Report Numbers
E 1.99:ornl/tm--12397
ornl/tm--12397
Subject(s)
Other Subject(s)
Note
Published through SciTech Connect.
02/01/1994.
"ornl/tm--12397"
"DE94009499"
Wood, R.J.; Hardy, J.H.; King, J.F.; Montgomery, B.H.; Yoder, G.L.; Felde, D.K.; McFee, M.T.; Siman-Tov, M.; Shourbaji, A.A.; Pawel, R.E.; Power, B.H.; Farquharson, G.
Type of Report and Period Covered Note
Topical; 02/01/1994 - 02/01/1994
Funding Information
AC05-84OR21400
View MARC record | catkey: 14365140